The present invention relates to a process for the recovery of uranium values in an extractive reprocessing process for irradiated nuclear fuels.
Until now, in order to recycle irradiated nuclear fuels, nuclear reactor fuel elements were dissolved, for example, in nitric acid, and the uranium separated by liquid/liquid extraction, as, for example, in the Purex process, or by amine extraction, or by column chromatography separation operations, and reprocessed in a nitric acid medium.
The nitric acid recycling of nuclear fuels constituted mainly of UO.sub.2, especially the Purex process, is a reliable process that has been known for a long time. After reaching the pre-determined length of operation or the desired burn-up, respectively, the fuel elements to be replaced are removed from the nuclear reactor and submitted, for example, to a one to three year storage for the cooling of the shorter lived fission products. Only after this storage duration are the fuel elements transported to the reprocessing installation and there divided into relatively small pieces, from which the remaining fission materials and the resultant fission products, etc., are dissolved out with strong nitric acid. The aqueous fuel solution thereby obtained is then diluted and fed into the first column of the first extraction cycle of the process. In the first extraction column, in counter current to the aqueous fuel solution, an organic extraction solution generally comprised of an organic extractant agent and an organic diluent agent is fed to extract or convert to an organic phase the fission materials uranium and plutonium, as well as smaller amounts of other actinides and small amounts of fission products. The aqueous, nitric acid run-off from the extraction column, now containing only very small amounts of uranium and plutonium, contains the main amount of fission products, corrosion products, etc., and represents a highly radioactive waste solution. After washing the organic phase with diluted nitric acid, the plutonium is treated with an aqueous stripping solution and transferred with simultaneous reduction of the oxidation state of the plutonium selectively from the organic phase into the aqueous phase. Then, the uranium still remaining in the organic phase (the main amount of the fission materials) is likewise transferred into an aqueous stripping solution. The aqueous solutions of uranium and plutonium are now further processed separately, for example, by means of two further purification cycles each, in order to be decontaminated as thoroughly as possible from the fission products as well.
Although this process method dominates for long irradiated and relatively long cooled fuel elements and can be safely carried out within suitable process conditions, it does have several disadvantages. For example, additional aqueous waste streams are obtained from the first extraction column at different places in addition to the highly radioactive waste solution, which contain radioactive fission products, etc. These waste streams must be concentrated and led either to further processing or to solidification. Moreover, small amounts of fission material can escape from the product streams of the process by the formation of degradation products from the extraction agents, whereby the degradation products form strong bonds with small amounts of fission materials and reach the aqueous waste streams.